Mechanical Properties of Ductile Cast Iron and Cast Steel for Intermediate Level Waste Transport Containers

Abstract

UK Nirex Ltd is developing Type B re-usable shielded transport containers (RSTCs) in a range of shielding thicknesses to transport intermediate level radioactive waste (ILW) to a deep repository. The designs are of an essentially monolithic construction and rely principally on the plastic flow of their material to absorb the energies involved in impact events. Nirex has investigated the feasibility of manufacturing the RSTCs from ductile cast iron (DCI) or cast steel instead of from forgings, since this would bring advantages of reduced manufacturing time and costs. However, cast materials are perceived to lack toughness and ductility and it is necessary to show that sufficient fracture toughness can be obtained to preclude brittle failure modes, particularly at low temperatures. The mechanical testing carried out as part of that programme is described. It shows how the measured properties have been used to demonstrate avoidance of brittle fracture and provide input to computer modelling of the drop tests.     

Testing of the Sealing Arrangements for the Nirex Re-Usable Shielded Transport Containers

Abstract

As part of its responsibility for the development of a deep repository for intermediate and low level radioactive waste, UK Nirex Ltd is developing a range of Type B re-usable shielded transport containers (RSTCs). A testing programme has been carried out on two alternative concepts for the RSTC sealing arrangements over the temperature range ?40^°C to 200^°C. For each sealing system, a test rig was developed to measure the performance under simulated normal and accident conditions of transport. The elastomer O-rings used for some of the tests had been irradiated to the maximum dose they might receive in normal transport. The performance of both sealing systems was good and it is concluded that either concept would meet the specified leakage criteria over the full temperature range under both normal and accident conditions of transport. However, further testing is required to confirm the performance of Concept N under accident conditions.     

Performance of the IAEA transport regulations in controlling doses and risks from a large-scale radioactive waste transport system

Abstract

The role of United Kingdom Nirex Ltd is to provide the UK with safe, environmentally sound and publicly acceptable options for the long-term management of radioactive materials generated by the UK's commercial, medical, research and defence activities. An important part of this role is to set standards and specifications for waste packaging. Waste producers in the UK are currently developing processes for packaging many different types of intermediate-level waste (ILW), and also those forms of low-level waste that will require similar management to ILW. When packaging processes are at the proposal stage, the waste producers consult Nirex about the suitability of the resulting packages for all future aspects of waste management. The response that Nirex provides is based on detailed assessments of the proposed packages, their compliance with Nirex standards and specifications, and their predicted performance through the successive phases of waste management. One of those phases is transport through the public domain. This paper draws on experience gained from more than 200 separate transport safety assessments which have cumulatively covered a wide range of waste types, waste packages and transport packages.     

NIREX Transport Activities since Rejection of the Rock Characterisation Facility Planning Application

Abstract

The application by Nirex to build an underground Rock Characterisation Facility, was rejected by the Secretary of State on 17 March 1997. This decision has caused Nirex to reconsider its forward programme for the development of an underground disposal facility for intermediate level waste. This paper describes the transport related activities being undertaken by Nirex since that date.     

Development of the Nirex Generic Transport Safety Assessment to Assist in the Provision Advice of Waste Packaging

Abstract

The current Nirex mission is to provide the United Kingdom with safe, environmentally sound and publicly acceptable options for the long-term management of radioactive materials. As part of this role, Nirex has developed a phased deep geological disposal concept which is defined by six ‘generic documents’ that describe systems, processes and safety assessments that are not specific to anyone location or geology. These generic documents give access to detailed information about the ideas and approaches that underpin the phased disposal concept, and have been published with an invitation to enter into dialogue with Nirex regarding these issues. The generic documents identify the requirements for an integrated transport system that would be necessary for the management of the intermediate-level (ILW) and low-level (LLW) wastes within Nirex's remit — the so-called reference case volume. This has involved Nirex in the development of transport hardware and associated safety reports and modelling and assessment tools for transport system logistics and system safety. Although the phased disposal concept is only one option for the long-term management of waste, the integrated transport system and associated modelling tools are likely to be of equal relevance to other options. The safety assessment of the generic transport operation for the movement of ILW and LLW waste from waste producers' sites to a future radioactive waste disposal facility is described in one of the generic documents — the generic transport safety assessment (GTSA). The GTSA demonstrates that the transport operation is compliant with Nirex safety principles, and that the nuclear and non-nuclear risks to the public and workers from routine transport and from accidents are acceptable. This paper describes the types of risk that are calculated, and discusses the data requirements and calculation methodology. The verification and validation methodology is outlined, together with a discussion of the results and a comparison of the risks with the Nirex dose and risk targets. In addition, this paper also describes how the methodology of the GTSA has been developed into an innovative software tool, TranSAT, which is routinely used as part of the packaging waste advice service offered by Nirex.     

Development of Sealed Type B Transport Containers for the Transport of Intermediate Level Radioactive Waste

Abstract

UK Nirex Ltd have developed three possible concepts of sealed transport containers for the safe transport of immobilised intermediate level radioactive waste. Computer based finite element impact and thermal analysis has been carried out on each concept and compliance with both the IAEA regulatory requirements and specified Nirex design aims has been demonstrated. One single concept will be selected at a later date following further development and confirmatory testing to provide a fleet of transport containers.     

Design and licensing of UKAEA IP 2m box

Abstract

Croft Associates Ltd (Croft) was contracted by the United Kingdom Atomic Energy Authority (UKAEA), Winfrith in 2006 to design, manufacture, test and license an intermediate level waste industrial package transport packaging (UKAEA 2 m Box Design no. 3954) for decommissioning waste from the Dragon and steam generating heavy water reactors at Winfrith. The package had to meet the requirements of the existing UK. Nuclear Decommissioning Authority (NDA) (Nirex) 2 m Box specification. The UKAEA 2 m Box is intended for on-site storage at Winfrith and possibly another site for 50 years and subsequent transport to the NDA repository without further processing. The design of the box closure includes a leak testable sealing system which is capable of being replaced at a later date (e.g. before shipment), by the removal of the lid and replacement of seals remote from the box location. Initially the specification for the UKAEA 2 m Box called for rating for gross weight of 40 t, but this was raised to 50 t as it was realised that the efficient use of the capacity of the box would result in a 50 t gross weight. Two prototype 2 m Boxes were produced; one being provided with 100 mm of concrete shielding, and the other being provided without any shielding. The shielded box was filled with simulated waste and a concrete grout cap fitted and the resulting 50 t package was dropped onto the massive target at Winfrith from a height of 300 mm (this being the regulatory test height). The test showed that the box performed as expected; meeting all requirements. In particular: there was no loss of shielding and no effect on the containment system that would reduce its effectiveness.     

Probabilistic Safety Assessment for the Transport of Radioactive Waste to a UK Repository at Sellafield

Abstract

A study has been undertaken to provide a detailed understanding of the radiological and non-radiological risks associated with the transpott of radioactive waste from the sites at which waste is produced in the UK to a proposed deep repository at Sellafield, and to ensure that these risks meet the design targets specified by Nirex. The routine transport collective dose to members of the public was assessed to be 0.2 man.Sv per year, which is only about 0.004% of the natural background dose. Accident frequencies were calculated using event tree methodology. The radiological consequences of accidents were assessed using the probablistic computer code CONDOR. The risk expectation value was calculated to be 1.5 × 10^?5 ? 8.6 × 10^?6 latent cancer fatalities per year (depending on the transport mode scenario). These values are significantly lower than the corresponding prediciions for non-radiological accident fatality rates, 0.05 ? 0.035 fatalities per year. The radiological accident risk for the most exposed individual member of the public was assessed to be 5 × 10^?11 ? 1.7 × 10^?11 per year, very much less than the Nirex target of 5 × 10^?7 per year. Plots of societal risk were shown to lie in the region of ‘negligible risk’, as defined by the UK Health and Safety Commission for non-radioactive dangerous goods transport.     

The Planned Integrated Transport System for the Deep Repository in the UK

Abstract

UK Nirex Ltd is responsible for developing a deep repository for the safe disposal of intermediate and low level radioactive wastes (ILW and LLW), and is concentrating its investigations on Sellafield as a potential location. A key part of the repository development programme is a transport system to deliver packaged wastes from sites elsewhere in the UK. The transport system must be able to handle a range of standard waste packages, and all transport through the public domain must comply with the IAEA Transport Regulations. Two design concepts have been developed for re-usable shielded transport containers for ILW, which are predicted to withstand accidents at least as severe as the IAEA Type B test conditions. Assessment, testing and further development of both concepts continues, with a view to selecting one for quantity production. Nirex is working closely with various organisations to establish the optimum transport routes for a potential repository at Sellafield. The current policy is that rail transport shall be used wherever practical for the transport of waste to the repository, although some road transport may also be required; the company has assessed a range of options. A Probabilistic Safety Assessment of the proposed transport operations has predicted that the radiological risks are expected to be extremely low, reflecting the adequacy of the packaging concepts. In addition, Nirex has identified a suitable transport emergency plan to deal with any unforeseen events.     

Conceptual Design of Large Boxes for the Packaging of Intermediate and Low Level Radioactive Waste

Abstract

United Kingdom Nirex Limited (Nirex) is developing standard containers for the packaging of radioactive waste for disposal in a deep underground repository. Waste Package Specifications have been produced for each standard package to provide the essential link between waste package design and the design of the deep repository. Previous studies carried out by Nirex identified the dimensions and key features of standard boxes for decommissioning intermediate level waste and for low level waste: the 4 m ILW box, 4 m LLW box·and 2 m LLW box. Nirex has now produced conceptual designs for these boxes.     

The Development of a ‘Large liner’ For the Processing and Solidification Of Intermediate Level Wastes

Abstract

Magnox Electric have developed a ‘Large Liner’ for the processing and solidification of the majority of the slurry form intermediate level wastes which arise on its power station sites (operational or undergoing decommissioning). The Liner comprises a cylindrical vessel of approximate dimensions 1.7 m diameter by 1.2 m high and it has been developed to comply with the requirements of the current Nirex Package Specifications. It is manufactured from stainless steel, being of all welded construction, with a bolted lid, incorporating an ‘O’ ring elastomer seal. The Liner can be fitted with a paddle for the ‘in drum’ solidification of slurry from wastes. This paper describes the development of the Liner, outlining the main stages from initial concepts to the final testing and receipt of approval from Nirex, who have adopted the Liner as one of its ‘Standard’ waste containers.     

Boxes for the Transport and Disposal of Low Level and Decommissioning Intermediate Level Radioactive Wastes

Abstract

UK Nirex Ltd is developing specifications for standard containers for the packaging of low and intermediate level radioactive wastes for disposal in a deep underground repository. The methodology used for selection of the dimensions and payloads of boxes for LLW and decommissioning ILW are described, and design features for handling the boxes are also briefly discussed. It must be emphasised that the results of this study are as yet preliminary, and may change in the light of the evolving system design for waste packaging, storage, transport and disposal.     

Transport of Fresh MOX Fuel Assemblies for the MONJU Initial Core

Abstract

Transport of fresh MOX fuel assemblies for the prototype FBR MONJU initial core started in July 1992 and ended in March 1994. As many as 205 fresh MOX fuel assemblies (109 assemblies for an inner core, 91 assemblies for an outer core and 5 assemblies for testing) were transported in nine transport missions. The packaging for fuel assemblies, which has shielding and shock absorbing material inside, meets IAEA regulatory requirements for Type B(U) packaging including hypothetical accident conditions such as the 9 m drop test, fire test, etc. Moreover, this packaging design features such advanced technologies as high performance neutron shielding material and an automatic hold-down mechanism for the fuel assemblies. Every effort was made to carry out safe transport in conjunction with the cooperation of every competent organisation. This effort includes establishment of, the transport control centre, communication training, and accompanying the radiation monitoring expert. No transport accident occurred during the transport and all the transport missions were successfully completed on schedule.     

Letter to the Editor

Abstract

Transport of fresh MOX fuel assemblies for the prototype FBR MONJU initial core started in July 1992 and ended in March 1994. As many as 205 fresh MOX fuel assemblies (109 assemblies for an inner core, 91 assemblies for an outer core and 5 assemblies for testing) were transported in nine transport missions. The packaging for fuel assemblies, which has shielding and shock absorbing material inside, meets IAEA regulatory requirements for Type B(U) packaging including hypothetical accident conditions such as the 9 m drop test, fire test, etc. Moreover, this packaging design features such advanced technologies as high performance neutron shielding material and an automatic hold-down mechanism for the fuel assemblies. Every effort was made to carry out safe transport in conjunction with the cooperation of every competent organisation. This effort includes establishment of the transport control centre, communication training, and accompanying of the radiation monitoring expert. No transport accident occurred during the transport and all the transport missions were successfully completed on schedule.     

Nirex—the Complete Package

Abstract

Through continuous interaction with its customers, Nirex identified the need for a comprehensive range of waste containers, reflecting the variety of wastes and operational undertakings. The current range consists of five standard containers. Standardisation is introduced across all waste packages to enable the safe and efficient operation of future waste management facilities. The practical lessons learned during the development of standard containers are in turn reflected in the container design work that Nirex has undertaken. They are also fed into the advice given to customers during evaluation of waste packaging proposals.     

Description of the German Concept of Final Storage Cask

Abstract

The German storage concept for the direct final storage of spent fuel assemblies from LWR reactors is described. The final storage concept is designed in such a way that it encompasses the whole spectrum of fuel elements to be stored from German reactors, Le. U fuel assemblies and MOX fuel assemblies with a mean bumup of 55 GW.d.t^?1 heavy metal were considered. The further design requirements are defined in such a way that the cask concept satisfies the conditions for type B(U) transport, interim storage and fmal storage. The safe long-term containment of the activity is guaranteed by an inner cask welded leak-tight; the sufficient shielding and the transport packaging are ensured by a shielding cask.     

Passive Detection of Concealed Strontium-90 Radioisotope Thermoelectric Generators in Transport

Abstract

Many radioisotope thermoelectric generators (RTGs) used in civilian and military applications employ ⁹⁰Sr, a high-energy beta emitter. The possibility of indirect or deliberate contamination of the environment with this isotope is not negligible. Detection and identification of a concealed ⁹⁰Sr RTG is possible by monitoring the continuous-energy spectrum of the bremsstrahlung radiation emitted as the source's betaparticles interact with the active material inside the source and with the surrounding container and shielding material. This detection concept is verified by measuring the energy spectrum of the bremsstrahlung radiation emitted from a small-activity shielded ⁹⁰Sr source, using two modest energy-resolution detectors: a NaI(Tl) crystal and a BC-408 plastic scintillator. These measurements were in turn utilised to verify the results obtained from Monte Carlo simulations, to determine the detector response for a realistic RTG source. The minimum detection time was then estimated as a function of the activity of the source and thickness and type of surrounding shielding.     

Expansion of the capabilities of the GA-4 legal weight truck spent fuel shipping cask

Abstract

General Atomics has developed the model GA-4 legal weight truck spent fuel cask, a high-capacity cask for the transport of four pressurised water reactor (PWR) spent fuel assemblies, and obtained a certificate of compliance (CoC, No. 9226) in 1998 from the US Nuclear Regulatory Commission (NRC). The currently authorised contents for this CoC, however, are much more limiting than the actual capability of the GA-4 cask to transport spent PWR fuel assemblies. The purpose of this paper is to show how the authorised contents can be significantly expanded by additional analyses without any changes to the physical design of the package. Using burn-up credit as outlined in US NRC Interim Staff Guidance 8, Revision 2, the authorised contents can be significantly expanded by increasing the maximum enrichment as the burn-up increases. Use of burn-up credit eliminates most of the criticality imposed limits on authorised package contents, but shielding still limits the use of the cask for higher burn-up, short-cooled fuel. By reducing the number of assemblies transported (downloading) to two and using shielding inserts, even high-burn-up fuel with reasonable cooling times can be transported.     

Excellox 6 and 7 Irradiated Fuel Transport Flasks

Abstract

The Sellafield nuclear fuels reprocessing plant in Cumbria has been receiving Light Water Reactor (LWR) fuel from European and Japanese power stations since the early 1970s. Virtually all of this has been delivered to Sell afield in flasks of composite design. This design of flask comprises a thick lead liner surrounding the fuel cavity and providing gamma radiation shielding, structural strength being provided by an enclosing steel shell. The composite flask has proved to be safe and efficient in operation but is now meeting the limits of its potential due to the trend towards higher burn up fuel High burnup fuel emits increasing levels of neutron radiation which the composite flask was not designed to accommodate; thus the need for new designs of flasks to carry the fuels of the future. The Excellox 6 and Excellox 7 irradiated fuel transport flasks have been developed over a period of five years. They are of monolithic construction and have been designed to complement existing composite flask types. The Excellox 6 flask is designed to carry high bumup PWR fuel up to 5 m in length from Europe, whilst the Excellox 7 flask is shorter and can carry high burnup PWR and BWR fuels up to 4.5 m in length from both Europe and Japan. As a consequence of meeting Japanese regulatory requirements the shielding design of the Excellox 7 differs from the Excellox 6.